ML20215M142

Amend 23 to License NPF-29,changing Tech Specs for Operation W/New Exxon Fuel Assemblies Replacing Spent GE Fuel Assemblies in Core

ML20215M142
Person / Time
Site:	Grand Gulf
Issue date:	10/24/1986
From:	Butler W Office of Nuclear Reactor Regulation
To:
Shared Package
ML20215M146	List: ML20215M142 ML20215M150
References
TAC-61930, NUDOCS 8610300065
Download: ML20215M142 (35)

Text

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NUCLEAR REG L OR'r COMMISSION E

ivASHINGTON, D. C. 20555 MISSISSIPPI POWER & LIGHT COMPANY MIDDLE SOUTH ENERGY, INC.

SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION DOCKET NO. 50-416 GRAND GULF huCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 23 License No. NPF-29 1.

The Nuclear Regulatory Commission (the Commission) has found that A.

The application for amendment by Mississippi Power & Light Company, Middle South Energy, Inc., and South Mississippi Electric Power Association,(thelicensees)datedJuly 14, 1986 as amended August 15 September 4, and September 5, and supplemented October 3,1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regula-tions set forth in 10 CFR Chapter I; B.

The facility will operate in confonnity with the application, the provisions of the Act, and the roles and regulations of the Comission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety.of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-29 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appenuix and the Environmental Protection Plan contained in Appendix B, as revised l

through Amendment No.

23, are hereby incorporated into this license.

Mississippi Power & Light Company shall operate the facility in f

accordance with the Technical Specifications and the Environmental Protection Plan.

hu.I.,03OOO65861004 I

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' 3.

Thic license amendment is ef fective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

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Orthei signedly Walter R. Butler, Director BWR Project Directorate No. 4 Division of BWR Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: October 24, 1986 i

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. 3.

This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Walter R. Butler, Director BWR Project Directorate No. 4 Division of BWR Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: October 24, 1986

ATTACHMENT TO LICENSE AMENDMENT N0. 23' FACILITY OPERATING LICENSE N0 NPF-29 DOCKET NO. 50-41_6_

Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Overleaf page(s) provided to maintain document completeness.*

Remove Insert 1-1 1-1*

1-2 1-2 B 2-1 B 2-1*

B 2-la B 2-2 B 2-2 B 2-3 B 2-3 8 2-4 B 2-4 3/4 1-1 3/4 1-1*

3/4 1-2 3/4 1-2 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-2a 3/4 2-2a 3/4 2-2b 3/4 2-2b*

3/4 2-7 3/4 2-7 3/4 2-7a 3/4 4-lb 3/4 4-lb*

3/4 4-Ic 3/4 4-Ic B 3/4 1-1 B 3/4 1-1 B 3/4 1-2 B 3/4 1-2*

B 3/4 2-1 B 3/4 2-1 B 3/4 2-la B 3/4 2-2 B 3/4 2-2 B 3/4 2-3 B 3/4 2-3 8 3/4 2-4 B 3/4 2-4

i l 1 Remove Insert B 3/4 2-5 B 3/4 2-5*

B 3/4 2-6 B 3/4 2-6 B 3/4 2-6a B 3/4 2-7 B 3/4 2-7 B 3/4 3-7 B 3/4 3-7 8 3/4 3-8 B 3/4 3-8*

B 3/4 4-1 B 3/4 4-1 B 3/4 4-la B 3/4 4-la*

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1. 0 DEFINITIONS The following terms are defined so that uniform interpretation of these specifications may be achieved. The defined terms appear in capitalized type and shall be applicable throughout these Technical Specifications.

ACTION 1.1 ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions.

AVERAGE PLANAR EXPOSURE

1. 2 The AVERAGE PLANAR EXPOSURE shall be applicable to a specific planar height and is equal to the sum of the exposure of all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in j

the fuel bundle.

AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1.3 The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the sum of the LINEAR HEAT GENERATION RATES for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

CHANNEL CALIBRATION 1.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the l

chan..el output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL i

CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The i

CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

I CHANNEL CHECK 1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall. include, where possible, comparison of the channel indication and/or stat'us with other indi-cations and/or status derived from independent instrument channels measuring i

the same parameter.

CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be:

a.

Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips.

I b.

Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

l The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested.

GRAND GULF-UNIT 1 1-1

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DEFINITIONS j

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CORE ALTERATION i

1.7 CORE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources, incore instruments or reactivity controls within the reactor l

pressure vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude cospletion of the movement of a component to a safe conservative position.

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CRITICAL POWER RATIO 1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by 3pplication of the XN-3 correlation to cause l

some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

1 DOSE EQUIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those, listed in Table III of TID-14844, " Calculation of Distance Factors for Power l

and Test Reactor Sites."

f DRYWELL INTEGRITY l

l 1.10 DRYWELL INTEGRITY shall exist when:

(

a.

All drywell penetrations required to be closed during accident l

conditions are either:

f 1.

Capable of being closed by an OPERA 8LE drywell automatic isolation system, or i

k 2.

Closed by at least one manual valve, blind flange, or f

deactivated automatic valve secured in its closed position.

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except as provided in Table 3.6.4-1 of Specification 3.6.4.

i b.

The drywell equipment hatch is closed and sealed.

f l

c.

The drywell airlock is in compliance with the requirements of l

Specification 3.6.2.3.

I d.

The drywell leakage rates are within the limits of Specification l

3.6.2.2.

i e.

The suppression pool is in compliance with the requirements of

)

Specification 3.6.3.1.

r f.

The sealing mechanism associated with each drywell penetration; e.g., welds, bellows or 0-rings, is OPERA 8LE.

a GRAND GULF-UNIT 1 1-2 Amendeent No. 23 l

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i 2.1 SAFETY LIMITS BASES

2.0 INTRODUCTION

The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs.

Safety Limits are established to protect the integrity of these barriers during normal ~ plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage ~is not directly observable, a step-back approach is used to establish a Safety Limit for the MCPR. MCPR I

greater than the applicable Safety Limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The 1

fuel cladding is one of the physical barriers which separate the radioactive j

materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion i

or use related cracking may occur during the life of the cladding, fission

)

.. product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal 2

stresses which occur from reactor operation significantly above design condi-tions and the Limiting Safety System Settings. While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still i

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greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a

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margin to the conditions which would produce onset of transition boiling, MCPR of 1.0.

These conditions represent a significant departure from the condition

""" M intended by design for planned operation.

2.1.1 THERMAL POWER. Low Pressure or Low Flow I

The use of the GEXL correlation is not valid for all critical power calcula-tions at pressures below 785 psig or core flows less than 10% of rated flow.

Therefore, the fuel cladding integrity Safety Limit is established by other 3

means. This is done by establishing a limiting condition on core THERMAL POWER l

with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi. Analyses show that with a bundle flow of j

28 x 103 lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 108 lbs/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical l

power at this flow is approximately 3.35 MWt. With the design peaking factors, 1

this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER.

l Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

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i GRAND GULFUNIT 1 8 2-1 Amendment No.16 l

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2.1 SAFETY LIMITS BASES THERMAL POWER. Low Pressure or Low Flow (Continued)

The Exxon Nuclear Company (ENC) XN-3 critical power correlation is appli-cable to the mixed core beginning with cycle 2.

The applicable range of tha XN-3 correlation is for pressures above 585 psig and bundle mass flux greater than 0.25M1bs/hr-ft2 For low pressure and low flow conditions, a THERMAL POWER safety limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig and below 10% RATED CORE FLOW was justified for Grand Gulf cycle 1 operation based on ATLAS test data. Overall, because of the design thermal-hydraulic compatibility of the ENC 8x8 fuel design with the cycle 1 fuel, this i

justification and the associated low pressure and low flow limits remain appli-cable for future cycles of cores containing these fuel designs.

q With regard to the 1;w flow range, the core's bypass region will be flooded at any flow rate greater than 10% RATED CORE FLOW. With the bypass region f13Sded, the aswociated elevation head is sufficient to assure a bundle mass flux of greater than 0.25 M1bs/hr-fta for all fuel assemblies which can approach critical heat flux. Therefore, the XN-3 critical power correlation is appro-

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priato for flows greater than 10% RATED CORE FLOW.

hana, The low pressure range for cycle 1 was defined at 785 psig. Since the XN-3 correlation is applicable at any pressure greater than 585-psig, the cycle 1 low pressure boundary of 785 psig remains valid for the XN-3 correlation.

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l l GRAND GULF-UNIT 1 B 2-la Amendment No. 23 l

SAFETY LIMITS BASES 2.1.2 THERMAL POWER, Hiah Pressure and High Flow The onset of transition boiling results in a decrease in heat transfer from the clad, elevated clad temperature, and the possibility of clad failure.

However, the existence of critical power, or boiling transition, is not a di-rectly observable parameter in an operating reactor. Therefore, the margin to boiling transition is calculated from plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution. The mar-gin for each fuel assembly is characterized by the critical power ratio (CPR),

which is the ratio of the bundle power which would produce onset of transition boiling divided by the actual bundle power. The minimum value of this ratio for any bundle in the core is the minimum critical power ratio (MCPR).

t The Safety Limit MCPR assures sufficient conservatism such that, in the event of a sustained steady state operation at the ' CPR safety limit, at least A

99.9% of the fuel rods in the core would be expected to avoid boiling transi-tion. The margin between calculated boiling transition (MCPR = 1.00) and the Safety Limit MCPR is based on a detailed statirtical procedure which considers the uncertainties in monitoring the core operating state. Once specific uncer-l tainty included in the safety limit is the un:ertainty inherent in the XN-3 critical power correlation.

ENC report XN-NF-524(A), Rev. 1, " Exxon Nuclear

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Critical Power Methodology for Boiling Water Reactors," Nov. 1983, describes p

the methodology used in determining the Safety Limit MCPR.

The XN-3 critical power correlation is based on a significant body of g-practical test data, providing a high degree of assurance that the critical power as evaluated by the correlation is within a small percentage of the actual critical power being estimated. The assumed reactor conditions used in l

defining the safety limit introduce conservatism into the limit because bound-ing high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition. Still further con-servatism is induced by the tendency of the XN-3 correlation to overpredict the number of rods in boiling transition. These conservatisms and the inherent i

I accuracy of the XN-3 correlation provide assurance that during sustained opera-tion at the Safety Limit MCPR there would be essentially no transition boiling in the core.

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i GRAND GULF-UNIT 1 8 2-2 Amendment No. 23 l

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Bases Table B 2.1.2-1

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GRAND GULF-UNIT 1 B 2-3 Amendment No. 23

Bases Table B 2.1.2-2

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i GRAND GULF-UNIT 1 B 2-4 Amendment No. 23 i

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3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN 1

LIMITING CONDITION FOR OPERATION 3.1.1 The SHUTDOWN MARGIN shall be equal to or greater than:

0.38% delta k/k with the highest worth rod analytically determined, or a.

b.

0.28% delta k/k with the highest worth rod determined by test.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4 and 5.

t ACTION:

4 With the SHUTDOWN MARGIN less than specified:

a.

In CPERATIONAL CONDITION 1 or 2, reestablish the required SHUTDOWN MARGIN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in ~at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

In OPERATIONAL CONDITION 3 or 4, immediately verify all insertable control rods to be inserted and suspend all activities that could reduce the SHUTDOWN MARGIN.

In OPERATIONAL CONDITION 4, establish l

SECONDARY-CONTAINMENT. INTEGRITY within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

c.

In OPERATIONAL CONDITION 5, suspend CORE ALTERATIONS

• and other l

activities that could reduce the SHUTDOWN MARGIN and insert all insertable control rods within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Establish SECONDARY CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

SURVEILLANCE REQUIREMENTS e

4.1.1 The SHUTDOWN MARGIN shall be determined to be equa'l to or greater than specified at any time during the fuel cycle:

I By measurement, p.*ior to or during the first startup after each a.

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refueling.

b.

By measurement, within 500 MWD /T prior to the core average exposure at which the predicted SHUTDOWN MARGIN, including uncertainties and calculation biases, is equal to the specified limit.

c.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after detection of a withdrawn control rod that is immovable, as a result of excessive friction or mechanical interfer-ence, or is untrippable, except that the above required SHUTDOWN

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MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod.

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• Except movement of IRMs, SRMs or special moveable detectors.

GRAND GULF-UNIT 1 3/4 1-1 I

REACTIVITY CONTROL SYSTEMS 3/4.1.2 REACTIVITY ANOMALIES LIMITING CONDITION FOR OPERATION 3.1.2 The reactivity difference between the monitored core k,ff and the predicted core k,ff shall not exceed 1% delta k/k.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With the reactivity difference greater than 1% delta k/k:

l Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, perform an analysis to determi,ne and explain the a.

cause of the reactivity difference; operation may continue if the l

difference is explained and corrected.

b.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REQUIREMENTS 9

4.1.2 The reactivity difference between the monitored core k,ff and the predicted core k,ff shall be verified to be less than or equal to 1% delta k/,k:

During the first startup following CORE ALTERATIONS, and l

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b.

At least once per 1000 MWD /T during POWER OPERATION.

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i GRAND GULF-UNIT 1 3/4 1-2 Amendment No. 23 l

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-s 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1

LIMITING CONDITION FOR OPERATION 3.2.1 During two loop operation all AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figure 3.2.1-1 as multiplied by the smaller of either the flow-dependent MAPLHGR factor (MAPFAC ) of Figure 3.2.1-2, or the f

power-dependent MAPLHGR factor (MAPFAC ) of Figure 3.2.1-3.

p During single loop operation, the APLHGR for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits as determined below:

a) for fuel types 8CR210 and 8CR160 - the limit shown in Figure 3.2.1-1 as multiplied by the smaller of either MAPFAC, MAPFAC or 0.86; f

p and b) for fuel type XN-1 the limit determined in "a" above for fuel type 8CR210.

APPLICABILITY: OPERA TIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RhTED THERMAL POWER.

ACTION:

During two loop operation or single loop operation, with an APLHGR exceeding the' limits of Figure I.2.1-1, as corrected by the appropriate multiplication factor for each type of fuel, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 1

4.2.1 All APLHGRs shall be verified to be equal to or less than the required l

limits:

a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, j

b.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c.

Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL R00 PATTERN for APLHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

GRAND GULF-UNIT 1 3/4 2-1 Amendment No. 23 l

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AVERAGE PLANAR EXPOSURE (mwd /ST)

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FIGURE 3.2.1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) s VERSUS AVERAGE PLANAR EXPOSURE FOR CORE FUEL TYPES 8CR210, i

8CR160, ANO XN-1

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20 40 60 80 100 12 0 CORE Fl.0W (% RATED) F FIGURE 3.2.1-2 MAPFACf GRAND GULF-UNIT 1 3/4 2-2a AmendmentNo.23i

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1.1 1.0

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FOR P)70%;

0.8 f0R 25%(Ps40%;

DURING OE 1.00P OPERATION

{0RE FLOW F $50%

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OR 40%<PS 100% *,

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N FOR 25%I P $40%;

C.S CORE FLOW F > S0%

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tO 40 60 80 10 0 120 CORE THERMAL POWER (% RATED)P FIGURE 3.2.1-3 MAPFAC P

GRAND GULF-UNIT 1 3/4 2-2b Amendment No.16 r

POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) for GE fuel shall not exceed l

13.4 kw/ft.

The LINEAR HEAT GENERATION RATE (LHGR) for ENC fuel shall not exceed the limits shown in Figure 3.2.4-1.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

With the LHGR of any GE fuel rod exceeding the 13.4 Kw/ft limit or with the LHGR of any ENC fuel rod exceeding the limit of Figure 3.2.4-1, initiate i

corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL. POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.4 LHGR's of both GE fuel and ENC fuel shall be determined to be equal to or less than their allowable limits:

4 a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, e

b.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, j

c.

Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating on a LIMITING CONTROL ROD PATTERN for LHGR, and d.

The provisions of Specification 4.0.4 are not applicable.

1 GRAND GULF-UNIT 1 3/4 2-7 Amendment No. 23 l

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REACTOR COOLANT SYSTEM i

SURVEILLANCE REQUIREMENTS (Continued)

The differential temperature requirements 4.4.1.1.5.b and c do not apply when the loop not in operation is isolated from the reactor pressure vessel.

i 4.4.1.1.6 The limits and setpoints of Specifications 2.2.1, 3.2.1, and 3.3.6 shall be verified to be within the appropriate limits within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of an operational change to either one or two loops operating.

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3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made sub-critical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition ~.

i Since core reactivity values will vary through core life as a function of fuel depletion and poison burnup, the demonstration of SHUTDOWN MARGIN will be performed in the cold, xenon-free condition and shall show the core to be subcritical by at least R + 0.38% delta k/k or R + 0.28% delta k/k, as appro-priate.

The value of R in units of % delta k/k is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated beginning-of-life core reactivity. The value of R aust be positive or zero and must be determined for each fuel loading cycle.

Two different values are supplied in the Limiting Condition for Operation to provide for the different methods of demonstration of the SHUTDOWN MARGIN.

The highest worth rod may be determined analytically or by test. The SHUTDOWN MARGIN is demonstrated by an insequence control rod withdrawal at the beginning of life fuel cycle conditions, and, if necessary, at any future time in the cycle if the first demonstration indicates that the required margin could be reduced as a function of exposure. Observation of subcriticality in this condi-tion assures subcriticality with the most reactive control rod fully withdrawn.

. This reactivity characteristic has been a basic assumption in the analysis of plant performance and can be best demonstrated at the time of fuel loading, but the margin must also be determined anytime a control rod is incapable of insertion.

3/4.1.2 REACTIVITY AN0MALIES Since the SHUTDOWN MARGIN requirement is small, a careful check on actual reactor conditions compared to the predicted conditions is necessary. Any changes in reactivity from that of the predicted core k,ff can be determined j

from the monitored core k,ff using the core monitoring system. In the absence of any deviation in plant operating conditions or reactivity anomaly, these l

values should be essentially equal since the calculational methodologies are consistent. The predicted core k,ff is calculated by a 30 core simulation i

code as a function of cycle exposure. This is performed for projected or anticipated reactor operating states / conditions throughout the cycle and is usually done prior to cycle operation. The monitored core k,ff is that calcu-lated by the core monitoring system for actual plant conditions.

A deviation in reactivity of more than 15 from that predicted is larger

[

than expected for normal operation, and therefore, should be thoroughly t

evaluated.

I GRAND GULF-UNIT 1 B 3/4 1-1 Amendment No. 23 l

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REACTIVITY CONTROL SYSTEMS 1

I BASES 3/4.1.3 CONTROL R005 The specifications of this section ensure that (1) the minimum SHUTDOWN MARGIN is maintained, (2) the control rod insertion times are consistent with' those used in the accident, non-accident and transient analyses, and (3) the potentia! of facts of the rod drop accident and rod withdrawal Orror' event are limited. The ACTION statements permit variations from the basic requirements but at the same time impose more restrictive crit.eria for continued operation.

A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum. The requirements for the various scram time measurements ensure that any indication of systematic prob- -

less with rod drivss will be investigated on a timely basis.

Damage within the control rod drive mechanism coulti be a generic problem, therefore with a control rod immovable because of excessive friction or mechani-cal interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.

Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the nonfully-inserted position are'

+

consistent with the SHUTDOWN MARGIN requirements.

The number of control rods permitted to be inoperable but trfppable could be more than the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must t;e shut down for investigation and resolution of the problem.

i The control rod system is designed to bring the reactor suberitical at a rate fast enough to prevent the MCPR from becoming less than the Safety Limit l

i during the limiting power transient analyzed in Section 15.4 of the FSAR. This analysis shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the specifications, provide the required protection and MCPR remains greater than the Safety Limit. The l

cccurrence of scram times longer than those specified should be viewed as an indication of a systematic problem with the rod drives and therefore the sur-l vaillance interval is reduced in order to prevent operation of the reactor for lcng periods of time with a potentially serious _ problem.

l.

The scram discharge volume is required to be OPERA 8LE so that it will be available when needed to accept discharge water from the control rods during a reactor scram and will isolate the reacto: coolant system from the containment l

when required.

Control rods with inoperable accumulators are declared inoperable and Spec-ification 3.1.3.1 then applies. This prevents a pattern of inoperable accumu-

~

i lators that would result in less reactivity insertion on a scram than has been cnalyzed even though control rods with inoperable accumulators may sti.11 be slowly scrammed via reactor pressure or inserted with normal drive water pres-sure. Operability of the accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of l;

the reactor.

GRAND GULF-UNIT 1 8 3/4 1-2 Amendment No.16 l

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.u 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temper-ature folloving the postulated design basis loss-of-coolant accident will not exceed the 2200*F limit specified in 10 CFR 50.46, 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50.46.

The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all i

the rods of a fuel assembly at any axial. location and is dependent only secondar-ily on the rcd to rod power distribution within an assembly. The peak clad tem-perature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification. This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor. The Technical Speci-fication AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR of the highest powered rod divided by its local peaking factor. The Maximum Aver-age Planar Linear Heat Generation Rate (MAPLHGR) limits of Figure 3.2.1-1 are multiplied by the smaller of either the flow dependent MAPLHGR factor (MAPFAC )

l 7

or the power dependent MAPLHGR factor (MAPFAC ) corresponding to existing core p

flow and power state to assure the adherence to fuel mechanical design bases during the most limiting transient. The maximum factor (MAPFAC) for single j

loop operation is 0.86.

For single-loop operation with ENC 8x8 fuel, a MAPLHGR limit corresponding to the product of the highest enriched GE fuel MAPLHGR, and the appropriate MAPFAC, can be conservatively used, provided that the average planar exposure is limited to 25,000 MWD /ST.

MAPFAC 's are determined using the three-dimensional BWR simulator code to f

analyze slow flow runout transients. Two curves for each fuel vendor are pro-vided for use based on the existing setting of the core flow limiter in the Recirculatinn Flow Control System. The curve representative of a maximum core flow limit if 107.0% is more restrictive due to the larger potential flow runout transient.

MAPFAC 's are generated using the same data base as the MCPR to protect p

p the core from plant transients other than core flow increases.

The daily requirement for calculating APLHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribu-tion shifts are very slow when there have not been significant power or control GRAND GULF-UNIT 1 B 3/4 2-1 Amendment No. 23 l

3/4.2 POWER DISTRIBUTION LIMITS BASES AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Ccatinued) rod changes. The requirement to calculate APLHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shifts while still allotting time for the power distribution to stabilize. The requirement for calculating APLHGR after initially determining a LIMITING CONTROL R0D PATTERN exists ensures that APLHGR will be known following a change in THERMAL POWER or power shape, that could place operation exceeding a thermal limit.

The calculational procedure used to establish the APLHGR limits is based on a loss-of-coolant accident analysis. The analysis was performed using calculational models which are consistent with the requirements of Appendix K to 10 CFR 50. These models are described in references 1, 6, and 8.

3/4.2.2 [ DELETED]

f GRAND GULF-UNIT 1 B 3/4 2-2 Amendment No. 23 l i

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Bases Table B 3.2.1-1

[ DELETED]

GRAND GULF-UNIT 1 B 3/4 2-3 Amendment No. 23 l

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POWER DISTRIBUTION LIMITS I

BASES 1

3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel clad-ding integrity Safety Limit MCPR, and an analysis of abnormal operational tran-1 For any abnormal operating transient analysis evaluation with the sients.

)

initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been snalyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant The limiting transient yields the largest delta CPR.

temperature decrease.

P c-added to the Safety Limit MCPR, the required operating limit MCPR of The power-flow map of Figure B 3/4 2.3-1 4t-+Mcu. ion 3.2.3 is obtained.

defines the analytical basis for generation of the MCPR operating limits.

The evaluation of a given transient begins with the system initial para-meters shown in FSAR Table 15.0-2 and in Table 15.C.3-1 of Reference 5 that The' eval-are input to a GE-core dynamic behavior transient computer program.

uation of transients during operation in the MEOD begins with the system initial parameters shown in Tables 15.D.4.2 and 3 of Reference 7.

The outputs l

of this program alcng with the initial MCPR fore the input for further analyses of the thermally limiting bundle. The principal result of this l

evaluation is the reduction in MCPR caused by the transient.

and MCPR is to define operating limits at other The purpose of the MCPRg p

than rated core flow and power conditions.

i*

The MCPR s are established to protect the core from inadvertent core flow f

The ref-increases such that the 99.9% MCPR limit requirement can be assured.

is a hypothesized erence core flow increase event used to establish the MCPRf l

slow flow runout to maximum, that does not result in a scram from neutron flux overshoot exceeding the APRM neutron flux-high level (Table 2.2.1-1 item 2).

!l The maximum runout flow value is dependent on the existing setting of the core flow limiter in the Recirculation Flow Control System. Two flow rates have been considered, 102.5% core flow and 107.0% core flow (for Increased Core Flow oper-curves are generated from a series of steady li ation). With this basis, the MCPRf state core thermal hydraulic calculations performed at several core power and flow conditions along the steepest flow control line. In the actual calculations a conservative highly steep generic representation of the 105% steam flow rod-line flow control line has been used. Assumptions used in the original calcula-j tions of this generic flow control lir.c were consistent with a slow flow increase I

transient duration of several minutes: (a) the plant heat balance was assumed l

,4 GRAND GULF-UNIT 1 B 3/4 2-4 Amendment No. 23 l

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_ POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued) to be in equilibrium, and (b) core xenon concentration was assumed to be constant The generic flow control line is used to define several core power / flow states at which to perform steady-state core thermal-hydraulic evaluations.

The first state analyzed corresponded to the maximum core power at maximum core flow (either 102.5% for Rated Core Flow operation or 107% of rated for Increase Core Flow operation) after the flow runout. Several evaluations were performed at this state iterating on the normalized core power distribution input until the limiting bundle MCPR just exceeded the safety limit Specifica-tion (2.1.2).

Next, similar calculations of core MCPR performance were deter-mined at other power / flow conditions on the generic flow control line, assuming the same normalized core power distribution. The result is a definition of the f performance requirement such that a flow increase event to the maximum MCPR flow will not violate the safety limit. (The assumption of constant power dis-tribution during the runout power increase has been shown to be conservative.

Increased negative reactivity feedback in the high power limiting bundle due to doppler and voids would reduce the limiting bundle relative power in an actual i

runout.)

The MCPR, is established to protect the core from plant transients other l

than core flow increase including the localized rod withdrawal error event.

Core power dependent setpoints are incorporated (incremental control rod with-drawai ifmits) in the Rod Withdrawal Limiter (RWL) System Specification (3.3.6).

l These setpoints allow greater control rod withdrawal at lower core powers where I

However, the increased rod withdrawal requires core thermal margins are large.

higher initial MCPR's to assure the MCPR safety limit Specification (2.1.2) is not violated. The analyses that establish the power dependent MCPR require-ments that support the RWL system are presented in GESSAR II, Appendix 158.

For core power below 40% of RATED THERMAL POWER, where the EOC-RPT and the reactor scrams on turbine stop valve closure and turbine control valve fast closure are bypassed, separate sets of MCPR, limits are provided for high and low core flows to account for the significant sensitivity to initial core

?-

flows. For core power above 40% of RATED THERMAL POWER, bounding power-dependent MCPR limits were developed. The abnormal operating transients The current analyzed for single loop operation are discussed in Reference 5.limitshavebeenva MCPR limits were found to be bounding. These MCPR p

for use during Cycle 2.

No change to the MCPR operating limit is required for l p

single loop operation.

At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, l

the reactor will be operating at minimum recirculation pump speed and the modera-I tor void content will be very small. For all designated control rod patterns t

which may be employed at this point, operating plant experience indicates that l

the resulting MCPR value is in excess of requirements by a considerable margin.

c l ;

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GRAND GULF-UNIT 1 B 3/4 2-6 Amendment No. 23

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i POWER DISTRIBUTION LIMITS I

BASES MINIMUM CRITICAL POWER RATIO (Continued)

During initial start-up testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum recirculation pump speed.

The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily require'Aent for calculating MCPR when THERMAL POWER is greater than or squal to 25% of RATED THERMAL POWER is sufficient since power distribution sh*fts are very slow when there have not been significant power or control rod changes. The requirement to calculate MCPR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shifts while still allotting time for the power distribution to stabilize. The requirement for calculating MCPR after initially determining a LIMITING CONTROL ROD PATTERN exists ensures that MCPR will be known following a change in THERMAL POWER or power shape, that could place operation exceeding a thermal limit.

3/4.2.4 LINEAR HEAT GENERATION RATE i

This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than the design linear heat generation even if fuel pellet densification is postulated.

The daily requirement for calculating LHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distri-bution shifts are very slow when there have not been significant power or control rod changes. The requirement to calculate LHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shifts while still allotting time for the power distribution to stabilize. The requirement for calculating LHGR after initially determining a LIMITING CONTROL R0D PATTERN exists ensures that LNGR will be known following a change in THERMAL POWER or power shape that could place operation exceeding a thermal limit.

References:

1.

General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566, November 1975.

2.

[ DELETED]

3.

[ DELETED]

4.

[ DELETED) 5.

GGNS Reactor Performance Improvement Program, Single Loop Cperation Analysis, General Electric Final Report, February 1986.

6.

General Electric Ccapany Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, Amendment 2, One Recircula-tion Loop Out-of-Service, NED0-20566-2, Revision 1, July 1978.

7.

General Electric Company, " Maximum Extended Operating Domain Analysis," March 1986.

8.

XN-NF-80-19(A), Volume 2 " Exxon Nuclear Methodology for Boiling Water Reactors: EXEM BWR ECCS Evaluation Model," Exxon Nuclear Company, September 1982.

GRAND GULF-UNIT 1 8 3/4 2-7 Amendment No. 23 l

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l INSTRUMENTATION l

BASES 3/4.3.9 TURBINE OVERSPEED PROTECTION This specification is provided to ensure that the turbine overspeed protection instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed. Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety-related components, equipment or structures.

3/4.3.10 NEUTRON FLUX MONITORING INSTRUMENTATION This specification is to ensure that neutron flux limit cycle oscillations are detected and suppressed.

In order to identify a region of the operating map where surveillance should be performed, analyses were performed by Exxon Nuclear Company for the Grand Gulf reactor consistent with the USNRC approved methodology as described in XN-NF-691(P)(A) dated August 1984.

The surveillance region was established as that region for which the calculated decay ratio is greater than or equal to a value of 0.60 and less than 0.90, between the 39% and 45% flow line. The resulting region is illustrated in Figure 3.4.1.1-1 and is identified as Region I.

Region IV is restricted from operation. This is the region where either the calculated decay ratio is greater than 0.90 or flow is below the 39% flow l?ne.

In Region III, below the 80% rod line, and Region II, where the decay ratio is below 0.60, no detect and suppress surveillance activities are required.

Neutron flux noise limits are also established to ensure the early detec-tion of limit cycle oscillations. Typical APRM neutron flux noiu levels at up to 12% of rated power have been observed. These levels are easify bounded by values considered in the thermal / mechanical fuel design. Stabi'ity tests have shown that limit cycle oscillations result in peak-to peak magt stude of 5 to 10 times the typical values. Therefore, actions taken to supprer, flux oscillations exceeding three (3) times the typical value are sufficient to ensure early detection of limit cycl 9 oscillations. The specification.ine ludes the surveil-lance requirement to establish the requisite baseline noise data and prohibits operation in the region of potential instability if the app.opriate baseline data is unavailable.

GRAND GULF-UNIT 1 8 3/4 3-7 AmendmentNo.23l

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3/4.4 REACTOR COOLANT SYSTEM BASES i

3/4.4.1 RECIRCULATION SYSTEM Operation with one reactor core coolant recirculation loop inoperable has been evaluated and found to remain within design limits and safety margins pro-vided certain limits and setpoints are modified.

The "GGNS Single Loop Opera-tion Analysis" identified the fuel cladding integrity Safety Limit, MAPLHGR limit and APRM setpoint modifications necessary to maintain the same margin of safety for single loop operation as is available during two loop operation.

Additionally, loop flow limitations are established to. ensure vessel internal vibration remains within limits. A flow control mode restriction is also incorporated to reduce valve wear as a result of automatic flow control attempts and to ensure valve swings into the cavitation region do not occur.

An inoperable jet pump is not, in itself, a sufficient reason to declare' a recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable. Jet pump failure can be detected by monitoring jet pump per-formance on a prescribed schedule for significant degradation. During two loop operation, recirculation loop flow mismatch limits are in compliance with ECCS LOCA analysis design criteria. The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA. In cases where the mismatch limits cannot be maintained, continued operation is per-mitted with one loop in operation.

Figure 3.4.1.1-1 describes the boundaries of the detect and suppress region as discussed in bases 3/4.3.10 In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50*F of each other prior to startup of an idle loop. The loop temperature must also be within 50*F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles. Since the coolant in the bottom of the vessel is at a lower temperature than the coolant in the upper J

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regions of the core, undue stress on the vessel would result if the temperature difference was greater than 100*F. During single loop operation, the condi-tion may exist in which the coolant in the bottom head of the vessel is not circulating. These differential temperature criteria are also to be met prior to power or flow increases from this condition.

The recirculation flow control valves provide regulation of individual recirculation loop drive flows; which, in turn will vary the flow rate of coolant through the reactor core over a range c,onsistent with the rod pattern and recirculation pump speed. The recirculation flow control system consists of the electronic and hydraulic components necessary for the positioning of the two hydraulically actuated flow control valves. Solid state control logic will generate a flow control valve " motion inhibit" signal in response to any one of several hydraulic power unit or analog control circuit failure signals.

The " motion inhibit" signal causss hydraulic power unit shutdown and hydraulic isolation such that the flow control valve fails "as is."

This design feature insures that the flow control valves do not respond to potentially erroneous control signals.

GRAND GULF-UNIT 1 B 3/4 4-1 AmendmentNo.23l 4

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REACTOR COOLANT SYSTEM BASES e

3/4.4.1 RECIRCULATION SYSTEM (Continued)

Electronic limiters exist in the position control loop of each flow control valve to limit the flow control valve stroking rate to 1011% per second in the opening and closing directions on a control signal. failure. The analysis'of the recirculation flow control failures on increasing and decreasing flow are presented in Sections 15.3 and 15.4 of the FSAR respectively.

The required surveillance interval is adequate to ensure that the flow control valves remain OPERABLE and not so frequent as to cause excessive wear on the system component h.

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